Commercial reprocessing of nuclear reactor fuels results in high-level waste (HLW) which is stored in tanks. The waste must be eventually removed from the tanks and solidified in a form suitable for transportation to a federal repository for final disposition.
The HLW in the tanks typically contains plutonium and other transuranic (TRU) elements in two fractions: an alkaline waste (supernatant) solution, and a sludge layer of iron hydroxide and fission product solids. This alkaline waste solution is first treated to remove .sup.137 Cs by passing a water-diluted solution through a series of ion exchange columns containing aluminosilicate ion exchanger. The resulting effluent is mixed with concrete and stored as low-level waste (LLW).
The TRU content of alkaline supernatant waste stored in tanks at West Valley Nuclear Services Co. (hereinafter West Valley or WV) is approximately 49 nanocuries per gram (nCi/g), and after the above described treatment the content is about 21 nCi/g. It is desired to reduce the TRU content to 10 nCi/g or less. A proposed treatment process for the sludge fraction of stored HLW involves washing with at least 4 batch contacts of water to remove excess sodium sulfate, then to mix with glass formers and vitrify into a final HLW form. The sludge wash solutions will also be processed by ion exchange to recover .sup.137 Cs.
In previous studies, however, it was found that 20% of the plutonium and 95% of the uranium in the sludge transferred to the water phase during washing. These values were substantially reduced by washing with pH 12.5 water, but the plutonium content still represented about 5% of the soluble plutonium, exceeding the required maximum actinide content in the concrete waste form (&lt;100 nCi/g of waste). A method for plutonium removal from sludge water washes was required.
The ion exchange process for recovery of .sup.137 Cs at WV has utilized the inorganic ion exchanger IE-96, a synthetic zeolite available from UOP. This material has been used for cesium recovery because of its high ion exchange capacity and decontamination factor (DF) values (&gt;40,000), and because it can be incorporated with glass formers and washed sludge to form borosilicate glass.
Previous studies by the inventors of the present invention showed that the IE-96 zeolite ion exchanger, when treated with a solution of titanium [4+] salt of isopropoxy [triethanolaminato] dissolved in isopropyl alcohol, extracts traces of plutonium from the WV alkaline wastes. However, use of this material has not been further considered based on safety considerations (volatile organics) and the need for a simple and reproducible large-scale preparation process.